Text

Amend 8 to License DPR-22,changing TS Re drywell-to-torus Vacuum Breaker requirements.Marked-up SE Encl W/Stated Changes
	ML20127J686
Person / Time
Site:	Monticello
Issue date:	02/26/1975
From:	Goller K; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20127J673	List: ML20127J671; ML20127J686; ML20127J699; ML20127J705; ML20127J723;
References
	NUDOCS 9211190397
	Download: ML20127J686 (15)
	v • d • e

__._ _ _._. ____.-

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i NORTHERN STATES P0t1ER COMPANY DOCKET NO. 50-263

?'.ONTICELLO NUCLEAR CENERATING PLANT t

AVENDMENT TO PROVISIONAL OPERATING LICENSE-1 Amendment No. 8 License No. DPR-22 1.

The Nucicar Reculatory Corsnission (the Commission) has found that:

A.

The application for amendment by the Northern States Power Company (the licensee) dated July 10, 1973, conolies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and rec,ulations set forth in 10 CFR Chapter I;.

B.

The facility vill operate-in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance-(i) that-the activitics authorized by this amendment can be conducted without endangerin;; the health and safety of the public, and (ii)' that such activities vill be conducted in compliance uith the-Co= mission's regulations; D.

The issuance of thin anendacnt will not-be inimical to the comon defense and security or to the health and safety of the public; and E. ~ Prior public notice-of this amendment is not required since the amendment does not involve a significant hazards consideration.

.2.

.iccordinaly,'the license is anendeu by.a channe to the Technical Specifications as indicated in.the attachment to thio license amendncnt and Paragraph 3.? of Facility Licenso.30. 3TR-22 is hereby-nnended to read as follows:

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forum AEC 318 (Rev. 9 53) AEcM 02-10

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Technical Specifications i

i The Technical Specifications' contained in j

Appendix A,as revised, are hereby incorporated-1 in the license. The. licensee shall operate J

i the facility in accordance with the Technical

-Specifications, as revised by issued changes thereto through Change No. 17."

3.

This license amendment is effective ns of the date of its I

issuance.

FOR THE IRICLEAR REGULATORY CO:ilSSIO i i

i Oridnal signd by:

Er! P. (4Her i

Karl R. Coller, Assistant Director Cor,0peratine Reactora Division of R6a'ctor Licensing 4

t Attachnent:

l Change No. 17 to the Technical Specifications a

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cate of Issuance: FEB 2 61975 i

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ATTACIDfENT TO LICD;SE AMENDMENT NO. 8 i

CHANCE NO.17 TO THE TECIINICAL SPECIFICATIONS s'

PROVISIONAL OPERATING LICENSI: NO. DPR-22 NORTHERN STATES POTTER COMPMiY M0!'TICrLLO NUCLEAR CCIERATIUC PLANT _

DOCKET NO. 50-263 2

i The Technical Specifications contained in Appendix A, attached to Provisional Operating License No. DPR-22, are hereby changed by replacing pages 147,153 and 158 with revised cages bearing the same nur.bers and additional pages 147A, 159A, 155B and 167A. Changed areas on the revised pages are reflected by marginal lines, J

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l SU 8t N A M I' 'Dr DATEW Form AEC-318 (Rev. 9-53) AECM O240 Tr u. s sovERNMENT PRfMTANG QFFICE31974.826 344

-t 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVT.ILLANCE REQUIREMENTS

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4.

Pressure Suppression Chamber-Drywell Vacuum 4.

Pressure Suppression Chamber-Drywell Vacuum Breakers Breakers Operability and full closure of the a.

When primary containment is required, all n.

dryuell-suppressica chamber vacuum breckers dryuell-suppression chamber vacuum shall be operable and positioned in the breakers shall be verified by performance closed position as indicated by the of the follo Ing:

position indication system, except during testing and except as specified in 3.7.A.

(1) Monthly each operable arywell-4.b and c belcw.

suppression chamber vacuum t

breaker shall be exerciced through b.

Any drywell-suppresulon chamber vacuum an opening-closing cyc1e.

breaker may be nonfully closed as j

indicated by the position indication and (2) Once each operating fuel cycle, alarm systems provided that drywell to 17 drywell to suppression chamber leake

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suppression chamber dif ferential pressure shall be demonstrated to be less decay does not exceed that shown on Figure than that equivalent to a one-inch diameter orifice and each vacuum 3.7.1.

breaker shall be visually inspected.

Up to two drywell-suppression chamber (Containment access re ;uired) c.

vacuum breakers may be inoperable provided that: (1) the vacuum breakers (3) Once each operating cycle, vacuum are determined to be fully closed and at breaker position indication and least one position alarm circuit is alarm systems shall be calibrat an operable or (2) the vacuum breaker is functienally tested.

(Containment secured in the closed position.

access required)

(4) Once ecch operating cycle, the vacuum breakers shall be tested to determine that the force required te open each valve from fully closed tc fully open does not e.xceed that equivalent to 0.5 psi acting on the suppression chamber face of the valve disc. (Containment access required) 147

-... ___._.-~-._._..._..._.__ _ __-_____ _. _... -

3.0 LLIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS I

d.

One position alarm circuit can be inoperable b.

When the position of any drywell-providing that the redundant position alarm suppression chamber vacuum breaker valve circuit is operable.

Both position alarm is indicated to be not fully closed at a 17 circuits may be inoperable for a period not tine when such closure is required, the to exceed seven days provided that all vacuum drywell to suppression chacher differential breakers are operable.

pressure decay shall be demonstrated to be less than that shown on Figure j

3.7.1 immediately and following any evidence of subsequent operation of 17 the inoperable valve until the inoperabic valve is restored to a normal condition, When both position alarm circuits are made c.

or found to be inoperable, the control panel indicator light status shall be recorded daily to detect changes in the vacuum breaker pcsition.

j S.

Oxygen Concentration 5.

Oxygen Concentration a.

After conpletion of.startup test program Whenever inerting is required, the primary l

and demonstration of plant electrical containment oxygen concentration shall be output, t he primary containment atmosphere measured and recorded on a weekly basis.

l shall be reduced to less than 5% oxygen with nitrogen gas whenever the reacter coolant 4

pressure is above 110 psig in the power operating cond ition, except as specified in i

3.7.'A.5.b.

.b.

Within the 24-hour period subsequent to l

piacing the reactor in the run mode i

follouing shutdown, the containment j

atmosphere oxygen concentration shall be reduced to less than 5% by weight, and raintaire:i

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in this condition. Deinecting may cornnence 24 h

hours prict to leuving t he run mode fcr a react or shutdcun.

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Bases Continued:

3.7 A.

primary Containment The purpose of the vacuum relief valves is to crialize the pressure between the dryvell and suppression chamber and between the suppression charber and reactor building during loss of coolant accident so that structural integrity of the containment is maintained.

The vacuum relief system between the pressure suppression chamber and reactor building consist of two 100% vacuum relief breaket. (2 parallel sets of 2 valves in series). Operation cf either system will naintain the pressure differential less than 1 ps:. The external design pressure is 2 pqig. One valve ray be out of service for repairs for a pn 1 of se an days. This period is based on the low probability that system redundancy would be requited duriro, this time.

If repairs cannot be cc pleted within seven days, the reactor coolact systen is brought to a conditica where vacuum relief is no lenger required.

The capacity of the ten (10) dryvell vacuum relief valves is sized to limit the pressure dif ferential between the suppression chanber and drywell during post-accident drywell cooling operations to less l

l than the design limit of 2 psi.

The relief valves are sized on the basis of the Bodega Bay prer wre i ^

suppression system tests.

Since they are in serien wit h the reactor building to suppression cha Aer vacuun relief valves pressure drop across these valves must be included in the evaluation of dr; cell negative ' pressures, even though there does not scpear to be a nechanism for causing negative pre esures in excess of the 2 psi design' pressure. With eight of the ten valves in service, the dif ferential pressure'across the valves for maximum flow cenJitions would increase. With this additional prc.ssure Containntnt drop the total differential pressure would still be less thaa the 2 psi design valve.

inter,rity would therefore not be impaired.

In additien to the. above considerations, postulated leakage through the vacuum breaker to the suppresr' n chamber air space could result in a partial bypass of pressure suppression in the event of a LOCA or a small or intermediate steam Icak. This ef fect could potentially result in execeding containment 17 design pressure. As a reruit of the leakage potential, the ccntainment response has been analyzed for a number of postulated conditions.

It was found that the maximum allowable bypass area for any postulated break size was equivalent to a six-inch dinneter opening.1 This bypass corresponds to a 1 Report on Torus to Drywell Yacuun Breaker Tests and Modificaticas for Monticello Maclear Generating plant. dated March 12, 1973, su'o it ted to Iir. D.,I. Shovholt, AEC-DL, from Mr. i. O. Mayer,!iSP 158

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3/4 inch opening of any one valve or.08 inch opening for all ten valves, measured at the bottom of f

the disc with the top of the disc at the seat.

The position indication system is designed to detect i

closure within 1/8 inch at the bottom of the disc.

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At each refueling outage and following any sigificant naintenance on the vacuum breaker valves,

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positive seating of the vacuum breakers will be verified by Icak test.

The leak test is conservatively

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f designed to demonstrate that leakage is less than that equivalent to Icakage through a one-inch orifice which is about 3% of the maximum allowabic.

This test-is planned to establish a baseline for

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valve performance at the start of each operating cycle and to ensure that vacuum breakers are maintained y

as nearly as possible to their design condition. This test is not planned to serve as a limiting j

condition for operation.

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s During reactor operation, an exercise test of the vacuum breakers will be conducted monthly. This 4

j test will verify that disc travel is unobstructed and will provide verification that the valves are closing fully through the position indication systen.

If one or more of the vacuum breakers do not I

sent fully as determined from the indicating system, a leak test will be conducted to verify that Icakage is within the maximum allowable.

Since the extreme lower limit of switch detection capability is approximately 1/16", the planned test is designed to strike a balance between the detection switch capability to verify closure and the maxinum allowable leak rate. A special test was performed to establish the basis for this limiting condition.

During the first refueling outage all ten vacuum 17 breakers were shinmed 1/16" open at the bottom of the disc. The bypass area associated with the shinning corresponded to 63% of the maximum allowable.1 The results of this test are shown in Figure l

3.7.1.

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I When a drywell-suppression chamber vacuum breaker valve is exercised through an opening-closing cycle, i

the position indicating lights at the remote test panels are designed to function as follows:.

r Full Closed 2 Green - On 2 Red

- Off L

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Intermediate Position 2 Green - Off

2. Red

- Off

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Full Open 2 Green - Off 2 Red

- On A

i, The remote test panel consists of a push button to actuate the air cylinder for testing, two red lights, t

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and two green lights for each of the ten valvcs. There are four independent limit switches on each' valve.. The two saitches contro11in;- the green lights are adjusted to provide an indication of disc.

opening of less than 1/3" at the bottom of the disc. These witches are also used to activate the valve position alarm circuits. The two switches controlling tha red ligher are adjusted to provide indication of the disc very ne,r the full open position.

The control roon alarm circuits are redundant and fail safe. This assures that no simple failure will defeat alarming to the control room when a valve is open beyond allouable and when power to the switches fails.

The alatn is needed to alert the operator that action cust be taFen to correct a nolfunction 17 or. to investigate possible changes in valve position status, or beth.

If the alarm cannot be cleared duc to' the inability to establish indication of closure of one or vore valves, additional testing is requitec The alarm systen allcus the operator to c'ake this evaluation on a timely basis. The frequency of the testing of the alarms is the same as that required for the position indication system.

Operability of a vacuum breaker valve and the four associated indicating light circuits shall be established by cycling the valve. The sequence of the indicating lights will be observed to be that previously described.

If both green light circuits are inoperable, the valve shall be considered inoperable and a pressure test is required irrcediately and upon indication of subsequent operation.

If both red-light circuita are inoperable, the valve shall be considered inoperable, houever, no pressure test is required if positive closure indication is present.

The 5!'or.ygen concentration minimizes the possibility of hydrogen ecmbustion following a lors of coolant accident.

Significant quantities of hydregen could be gencrated if the core cooling systems failed to sufficiently cool the core. The occurrence of primary system leakage following a rajor

[

refueling outage or other scheduled shutdown is mere probabic than the occurrence of the lomt of coolant accident upon which the specified oxygen concentration limit is based. Permitting cecess to the dryuell for Icak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety. Titu s, to preclude the passibility of starting the reactor and operatinn for exteticd periods of tiac with significant leaks in the prirm. f (Continued on page 159) 1588

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.I SAFETY EVALUs. TION BY THE DIVISION OF REACTOR L_I_ CENSING SUPPORTING AMENDMENT NO. 8 TO LICENSE DPP-72 (C11ANCE NO._17__TO THE TECHNICAL SPECIFICATIONp NORT'IEPJi STATES PO'.7Q C0!!PANY,

?t0NTICELLO Ul]_ CLEAR GENERATING PLANT DOCEET NO. _50.263_

4 TOR TO _DR_Y_1[ ELL VACUUit BREAKER LEAKAGE

/

j INTRODUCTION A lettor(1)from the Directorate of Licensina requested that tho

lorthern States Power Company (NSP) provide torus-to-dryvoll vacuuti brenhor denien and test infomation.

The letter also recuented:

(1) a reevaluation of breaker performance, (2) additional equipment.

and nystems and/cr nodification to the vacuum breakers, and (2) 4 technical specification channen or additions related to linitinr, conaitions of operation and surveillance.

Northern States Power Cotpany provided the requested infornation(2)within 60 days.

Daned on their response.(2) flSP later requested (3) a chanc,e to the

"'echnical %ecifications, Appendi:: A, of the "rnvisionni mratirr License, DPN 22 for the !!onticello Nucicar Cennrating, Plant. Notet(A) with NSP representatives on September 6,1973, to discuss details of the torus-to-drywell vacuum brenhor valve position indientora and alarn circuits and to review the adequacy of proposed Icah test requirements for the Technical Specifications. As a result of the meeting, NSP descy(bgl an additional circuit nodification(5) to the original proposal "'h such that torus-drywell vacuun breaker switchen i

vill cause an audible alarr u nell as ranel light indication in the concrol roon if any of the 10 vacuur. brealern onen.

Mditional chancen to the JSP nroposal, which lyvc evolved fren discunnionn with N".P representatives, have been incorporated.

These changes relate to surveillsace and test requirements, specification 4.7.A.4a (3) and (4), 4.7.A.4c and limiting conditions for operation, specification 3.7.A.4(d).

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2-l EVALUATION 1

The drywell suppression chamber (torus) vacuum breakers protect the drywell from damage by a drywell negative pressure differential i

that could result with most of the non-condensiblo gas collected in the torus above the suppression pool after the water vapor in the drywell condenses following a design basis loss-of-coolant accident.

Testo of vacuum breakers in come BWR plants revealed that some breakers were not in the fully closed position as they were designed to be during normal reactor operating conditions. A partly open vacuum breaker l

would permit steam to bypass the suppression pool following loss-of-l coolant accidents causing higher-than-design containment pressure.

In response to our request (1)

NKP presented the calculated drywell-to-torus leak rates that could 4 I L

-* : e., ror primary system break areas as large as the design bs'.ta :.t (t= 1 break.

The results l

showed the variation in allownbe 67)at ut em Lcus leakage with theprigarysystembreakarea.

to n)

+y system breaks greater than 0.3 ft the allowable drywell-to-e,ua leakage increases, i.e.

thedrywell-to-torysequivalentbypasaincreAsesfromabout0.2ft' to more than i ft For primary system breaks less than 0.3 f t 2, i

l the allowabic drywell-to-torus leakage is Icar, than 0.2 f t 2, ye have reviewed the calculational method and assumptions used by NSP and have concluded that there is sufficient conservatism in the

]

calculations.

Therefore,thecalculatedjrywell-to-torusbypass leakage equivalent to that from a 0.2 ft (6 inch diameter) equivalent orifice is a justifiable limit for the entire range of core coolant breaks up to the acsign basis accident.

if the vacuum breaker opens during normal operation, drywell-to torus leak rate tests must be performed to show that the leakage is not excessive.

NSP has demonstrated that all breaker discs wedged open 1/16 inch will resylt in less than the allowable leakage, i.e.

leakage through 0.2 ft - equivalent orifice. We have concluded that the requirements for leak tests during reactor operation proposed by NSP are adequate to assure a 37% margin to the 0.2 ft 2 equivalent orifice limit value (maximum allowable leak for continuous operation is i

63% of the calculated limit - 6" diameter opening). Figure 3.7.1 of the proposed technical specification change defines, therefore, an acceptable test limit for continued reactor operation. !!oreover, at least once during each fuel cycle, while the reactor is shut down and the containment is accessible, it must be demonstrated that the combined drywell-to-torus leakage is less than the Icakage through a-f

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equivalent 1-inch orifice (less than 3% of the calculated leakage limit).

At the' name time all vacuum breakers must be visually inspected to nasure proper valve disc seating. These and other proposed outveillarce wouirments (Proposed Technical Specification 4.7.4) we have concluded, further reduce the probability that containment pressure vill exesed design limien following any break in the cora coolant piping systems.

Uo have reviewed the change in the v[cuum breaker closure switch desien.

We agree that replacement of Snap Lock switches attached to the valva shaf t with "icro Switches located on the botto:n of the valve seat inproves sensitivity in detecting valve disc novement from the closed

' position. Two !!1cro Switches have bcon inntailed on cach valve and wired to separate valve position indicator panels in the reactor building and control room. These nodifications permit verification of valve closure and natisfy roquirements for synten reoundancy.

The maifications including the alarm, which sounds if a valve dine Icaves the seat _ represent a substantial improvement in torun-to-dryvell vacutua breaker perfornance reliability.

.ul of the proposed modificationc have been installed and the systems are now operational. Thc proposed technical specifications as nodified by MSp and tho 00 accff vill enhance raector W ety and should be enproved.

Althour,h tests have shown that leakar.e fron the valve dice shaft is nonligibic with the teflon packinn removed, we understand that NSP v111 ropack one of the vacuue bronher shnfta with new notorial durine the acheduled refueline outage in January 1975. The renntning valvo shafts may be repacked after cufficient operating tine has elapsed to confirm the adequacy of the new packing.

C0::CLUSION tased on calculated drywell-to-torus Icak rate limits provided by NSP for the npectrum of desinn hosin core coolant breaks, un hwa concluded that the limit curve for leah :csta durinr, reactor operation is acceptable.

'te _ nico have concluded that the Icak rate test to be performed at the end of each operating cycle vill reasonably asnure that the torus-to

'rvucil vacuun trenhero are aroperiv.1cated Secounc monoured leaFace

'ro. the drvuoll to the torus nunt 50 leaa than E of the cafety limit before returulw. to r? actor orcration. 1 c.

971 car;',in to t ntint" linit. Uc anece with !!Sp that:

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1.

redundant audible alarms, as now provided at Monticello, assure that a valve opening more than 1/16 inch will be l

detected promptly during normal plant operation.

t j

2.

redundant indicator panels will show which valve has opened

)

3.

surveillance and test requirements are adequate considering j

the vacuum breaker performance history, test results, and the recentfy completed vacuum breaker system modifications.

4 For these reasons we have concluded that the proposed changes, as modified with mutual consent, will enhance reactor safety by assuring that containment pressure following loss of core coolant accidents will not exceed the 62 psig design limit because of faulty torus-to-drywell vacuum j

breaker (s).

The steam released during blowdown will condenso in the j

suppression pool as described in the FSAR.

\\

We have concluded, based on the considerations discussed above,

{

that:

(1) because the change doco not involve a significant increase

/

in the probability or consequences of accidents previously considered t

l and does not involvo a significant decrease in a safety margin, the J

change does not involve a significant hazards consideration. (2) therc is reasonabic assurance that the hcalth and safety of the public will not be endangered by operation in the proposed manner, en4-(3) such activitics will be conducted in compliance with the Commission's regula-j i

tions and the issuance of this amendment will not be inimical to the l

commo1 def e se and securf ty or to the health and safety of the public7 e nv f ro n mfpq J m oo c-I-r n;[emc/r' W

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REFERENCES y-i

)

j (1)- Directorate of Licensing letter dated January 12, 1973.

Request for design and test information related to torus-drywell vacuum i

breakers, plus evaluation of improvements and submittal of new j

technical specifications.

1 i

(2)

NSP letter dated March 12, 1973 response to AEC letter dated January j

12, 1973.

Report on Torus to Drywell Vacuum Breakers Tests and Modificatio,ns for Monticello Nuclear Generating Plant."

(3) NSP letter dated July 10, 1973.

Request for changes to the Monticello i

Technical Specifications to improve torus-to-drywell vacuum breaker l

reliability, i

(4) Directorate of Licenning Memo to Files dated September 11, 1973.

j Minutes of September 6. 1973 meeting with NSP representatives to discuss Monticello Torus-to-Drywell Vacuum Breakers.

i-(5) NSP letter dated September 17, 1973.

"Further Vacuum Breaker j

Modifications."

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